Cladding for use in nuclear fuel rods for light water reactors functions to prevent fission products from being released from the fuel into the coolant/moderator and to prevent contact and chemical reactions between the fuel and the coolant/moderator. The cladding is required to have excellent mechanical properties and high corrosion resistance in the environment and for the conditions expected during reactor operations. Cladding is therefore required to have adequate corrosion resistance for the lifetime of the fuel rod for operation in water and steam at temperatures up to approximately 345.degree. C., adequate strength and creep behavior over the lifetime of the fuel rod, and typically have low parasitic neutron absorption for economic use of the fissionable fuel material.
Common cladding materials include zirconium, zirconium alloys, and stainless steel. Zirconium based alloys in which the major component is zirconium have been used in the cladding of nuclear fuel rods or elements for several decades. Two of the most commonly used zirconium alloys that have given satisfactory performance are Zircaloy 2 and Zircaloy 4 and are described in American Society for Testing and Materials standard B350-93(1993), Standard Specification For Zirconium and Zirconium Alloy Ingots For Nuclear Application, compositions R60802 and R60804, respectively. Zircaloy 2 (composition R60802) is composed of from 1.20 to 1.70 weight percent tin, 0.07 to 0.20 weight percent iron, 0.05 to 0.15 weight percent chromium, 0.03 to 0.08 weight percent nickel, where the iron plus chromium plus nickel content is from 0.18 to 0.38 weight percent, and the balance is zirconium plus impurities. Zircaloy 4 (composition R60804) is composed of from 1.20 to 1.70 weight percent tin, 0.18 to 0.24 weight percent iron, 0.07 to 0.13 weight percent chromium, where the iron plus chromium content is 0.28 to 0.37 weight percent, and the balance is zirconium plus impurities. The maximum impurities for Zircaloy 2 and Zircaloy 4 are given in the following table which is from Table 1 of the ASTM B350-93 Standard.
TABLE I ______________________________________ MAXIMUM IMPURITIES, WEIGHT % R 60802 R 60804 ______________________________________ Aluminum 0.0075 0.0075 Boron 0.00005 0.00005 Cadmium 0.00005 0.00005 Carbon 0.027 0.027 Cobalt 0.0020 0.0020 Copper 0.0050 0.0050 Hafnium 0.010 0.010 Hydrogen 0.0025 0.0025 Oxygen * * Magnesium 0.0020 0.0020 Manganese 0.0050 0.0050 Molybdenum 0.0050 0.0050 Nickel -- 0.0070 Niobium 0.010 0.010 Nitrogen 0.0065 0.0065 Silicon 0.012 0.0120 Tin -- -- Titanium 0.0050 0.0050 Tungsten 0.010 0.010 Uranium (Total) 0.00035 0.00035 ______________________________________ * When so specified in a purchase order, oxygen shall be determined and reported. Maximum or minimum permissible values, or both, shall be as specified.
Although these and other alloys have provided generally adequate performance, they possess some deficiencies that have prompted further analysis and research to find alternative materials for and alternative constructions of nuclear fuel rod cladding to single walled cladding comprised of a single metal or alloy (sometimes referred to as "through" wall cladding) which does not possess both optimum strength and resistance to corrosion. Alternative constructions to single or through wall cladding for use as nuclear fuel rod cladding includes two layer or multiple layer tubing. These types of cladding have (a) an outer layer of a highly corrosion resistant alloy and (b) an inner layer that provides the bulk of the mechanical strength of the cladding. Cladding of this type, sometimes referred to as duplex cladding, with an extra low tin Zircaloy-type outer layer (nominally 0.8 wt. % tin) and a Zircaloy-4 inner layer is currently in use for nuclear fuel rod cladding. Zircaloy-4 inner layer cladding with a thin outer layer (3 to 5 mil) of various other corrosion resistant alloys has been produced and tested in-reactor. An outer layer alloy containing 0.5 wt. % tin, 0.5 wt. % iron, balance zirconium, and another outer layer alloy containing 0.5 wt. % tin, 0.5 wt. % iron, 0.2 wt. % chromium, balance zirconium have each shown exceptional corrosion performance in a high temperature pressurized water reactor. Examples of multiple layered tubing constructions and alloys for nuclear fuel rods are discussed in U.S. Pat. Nos. 5,493,592; 4,963,316; 4,735,768, which are each hereby incorporated by reference.
With the higher burnups and longer in-reactor residence times that are being pursued and which, for largely economic reasons, continue to be increased, performance limits of commonly used alloys for nuclear fuel rod cladding are being reached. The corrosion resistance of the Zircaloys has been a major concern, especially in modern high coolant temperature pressurized water reactors that employ low leakage core loadings where the corrosion film on Zircaloy can build up to unacceptable levels for burnups around 50 to 60 MWd/kgU. In efforts to optimize the corrosion performance of the Zircaloys, through a reduction in the tin level, the strength and creep properties of the cladding material have thereby been diminished. For example, over the last decade the tin level of the Zircaloys used as cladding materials in nuclear fuel rods which was nominally held at approximately 1.55 wt. % has been lowered to a nominal level of approximately 1.30 wt. %. This reduction in the level of tin has resulted in substantially better corrosion performance specifically at higher burnups, but the reduction in tin has negatively impacted the mechanical properties of the cladding. Tin is a solute solution strengthening alloy element in Zircaloy and improves the strength and creep resistance of the alloy. However, lowering the tin level in Zircaloy reduces the resistance of the cladding to creepdown as well as the strength of the cladding.
In attempts to overcome the limitations in the higher burnup performance of the zirconium alloys and the Zircaloys, alloy development programs have been initiated and research and development continue to this date for zirconium alloys for use as a nuclear fuel rod cladding that would have a more favorable combination of corrosion resistance, high strength and creep resistance as well as a low neutron cross section.
An object of the present invention is to improve upon the nuclear fuel rod claddings produced to date by using (I) an alloy for the outer layer of a multiple layered cladding tube with exceptional in-reactor corrosion characteristics and in accordance with the present invention to utilize (II) a new alloy for the inner part of the cladding that is of substantially higher strength than Zircaloy-2 or Zircaloy-4, while maintaining low parasitic neutron absorption characteristics of the latter alloys.
By using such a higher strength alloy for an inner layer of a multiple layer cladding tube, the overall cladding tube wall thickness can be reduced while still meeting the mechanical design and performance criteria of the fuel rod. By being able to reduce the cladding wall thickness, the cladding weight per unit length of cladding can be reduced and the cost of a cladding tube of a given length is reduced since less material is needed for the production of the cladding. Furthermore, by being able to reduce the cladding wall thickness, improvements in fuel cycle costs resulting from a reduction in the parasitic thermal neutron absorption can be obtained since parasitic neutron absorption for cladding of a given composition is directly proportional to the cladding wall thickness. Alloying elements with a smaller thermal neutron cross section than currently employed tin or niobium additions can reduce the parasitic neutron absorption of the alloy even further and gain additional improvements in fuel cycle costs.
By using such a higher strength alloy for an inner layer of the multiple layer cladding, significant energy production cost savings can also be obtained by reducing cladding wall thickness and increasing fuel rod fissionable material weight which is achieved by being able to use larger diameter fuel pellets while maintaining a constant fuel rod outer diameter. For a given fuel rod design, the outer diameter of the cladding is primarily determined by thermal hydraulic considerations and therefore cannot readily be changed. Thin wall cladding can accommodate larger diameter fuel pellets than a thicker wall cladding of the same outside diameter. A larger diameter fuel pellet can have a lower uranium enrichment than a smaller diameter pellet to produce the same amount of energy. For slightly enriched uranium dioxide nuclear fuel, the lifetime energy production of a unit length of fuel rod is proportional to the total number of U.sup.235 atoms per unit length. Thus, for example, by using cladding with a 0.005 inch thinner wall than a thick wall design fuel rod containing 0.300 inch diameter pellets enriched to 4.00 wt. % U.sup.235, fuel pellets of 0.310 inch diameter may be used. The reduced U.sup.235 enrichment of these pellets would be ##EQU1## (where L is a unit length of fuel) to maintain approximately the same number of U.sup.235 atoms per unit length of fuel. Alternatively, by maintaining the same U.sup.235 enrichment and increasing the pellet diameter, the number of U.sup.235 atoms per unit length of fuel rod is increased and the lifetime energy production of a unit length of fuel would be increased as well. Either alternative would lead to reactor fuel cycle cost reductions by using relatively higher cost, but thin wall, multiple layer cladding compared to using thicker through wall Zircaloy cladding.